05000306/LER-2012-001, For Prairie Island Nuclear Generating Plant, Regarding Unit 2 Manual Reactor Trip Due to Feedwater Heater Hi Hi Alarm

For Prairie Island Nuclear Generating Plant, Regarding Unit 2 Manual Reactor Trip Due to Feedwater Heater Hi Hi Alarm

ML12111A172
Person / Time
Site:	Prairie Island
Issue date:	04/19/2012
From:	Schimmel M Xcel Energy, Northern States Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-12-032 LER-12-001-00
Download: ML12111A172 (5)

LER-2012-001, For Prairie Island Nuclear Generating Plant, Regarding Unit 2 Manual Reactor Trip Due to Feedwater Heater Hi Hi Alarm

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3062012001R00 - NRC Website

text

@ Xcel Energym April 19, 2012 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 L-PI-I 2-032 10 CFR 50.73 Prairie lsland Nuclear Generating Plant Unit 2 Docket: 50-306 Renewed License No. DPR-60 LER 50-306/2012-001-00, Unit 2 Manual Reactor Trip Due To Feedwater Heater Hi Hi Alarm Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, herewith encloses Licensee Event Report (LER) 50-3061201 2-001-00.

On February 21, 2012, Prairie lsland Nuclear Generating Plant (PINGP) Unit 2 was performing a normal shutdown in preparation for refueling outage 2R27. With Unit 2 at approximately 11.42% power, the reactor was manually tripped in accordance with the 21/22/23 Feedwater Heater Hi Hi alarm response procedure. There were no other unusual / not understood events associated with the shutdown.

The causal evaluation for the reported condition is in progress. A supplement to the LER will provide additional information from the casual evaluation.

Summarv of Commitments This letter contains no new commitments and no changes to existing commitments.

Mark A. Schimmel Site Vice President, Prairie lsland Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Prairie lsland Nuclear Generating Plant (PINGP), USNRC Resident Inspector, PINGP, USNRC Department of Commerce, State of Minnesota 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651 -388.1 121

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

LICENSEE EVENT REPORT (LER)

(See reverse for required number of digitstcharacters for each block)

APPROVED BY OM6 NO 3150-0104 EXPIRES 10/31/2013

, the NRC may not conduct or sponsor, and a person is not required to respond to, the lnformation colledlon

1. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 2

2. DOCKET NUMBER 05000 306

3. PAGE 1

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4. TITLE Unit 2 Manual Reactor Trip Due To Feedwater Heater Hi Hi Alarm

5. EVENT DATE

6. LER NUMBER SEQUENTIAL REV YEAR 1 NUMBER I NO 2012 - 001 - 00 MONTH 02

12. LICENSEE CONTACT FOR THlS LER

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check all that apply)

C]

20.2201(b)

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20.2203(a)(3)(i)

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50.73(a)(2)(i)(C)

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50.73(a)(2)(vii)

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20.2201(d)

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20.2203(a)(3)(ii)

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50.73(a)(2)(ii)(A)

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50,73(a)(2)(viii)(A)

[7 20.2203(a)(I)

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20.2203(a)(4)

[1 50.73(a)(2)(ii)(B)

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50,73(a)(2)(viii)(B)

[7 20.2203(a)(2)(i)

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50,36(c)(l)(i)(A)

[7 50.73(a)(2)(iii)

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50,73(a)(2)(ix)(A)

[1 20.2203(a)(2)(ii) 50,36(c)(l)(ii)(A) 50.73(a)(2)(iv)(A)

[1 50,73(a)(2)(x) 20.2203(a)(2)(iii)

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50.36(~)(2)

[1 50,73(a)(2)(v)(A) 73,71(a)(4)

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2OS2203(a)(2)(iv)

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5OS46(a)(3)(ii)

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50.73(a)(2)(v)(B)

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73.71(a)(5) 20.2203(a)(2)(v)

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50,73(a)(2)(i)(A)

[7 50.73(a)(2)(v)(C)

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OTHER C]

20.2203(a)(2)(vi)

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5OS73(a)(2)(i)(B)

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50.73(a)(2)(v)(D)

Spec~fy in Abstract below or in 9. OPERATING MODE Mode 1

10. POWER LEVEL 11.42%

DAY 21 NAME Sam J. DiPasquale, P.E.

YEAR 2012

7. REPORT DATE TELEPHONE NUMBER (Include Area Code) 651.388.11 21 x7350 MONTH 04

8. OTHER FACILITIES INVOLVED FACILITY NAME FACILITY NAME DAY 19 DOCKET NUMBER DOCKET NUMBER

CAUSE

YEAR 2012 SYSTEM

14. SUPPLEMENTAL REPORT EXPECTED 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE).

0 NO COMPONENT MANU-FACTURER ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On February 21, 2012, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was performing a normal shutdown in preparation for refueling outage 2R27. With Unit 2 at approximately 11.42%

power, the reactor was manually tripped in accordance with the 21/22/23 Feedwater Heater Hi Hi alarm response procedure. There were no other unusual / not understood events associated with the shutdown.

The Equipment Cause Evaluation (ECE) determined that the third stage low pressure FWH bypass line to the condenser is potentially restricting flow to the dump valve. The ECE also determined that the Moisture Separator Reheater Control valves are ramped closed earlier than necessary and are fully closed by approximately 20% reactor power. This causes excessive moisture in the extraction steam at low power and results in more water accumulating in the low pressure feedwater heaters.

NRC FORM 366 (10-2010)

REPORTABLE TO EPlX

CAUSE

YEAR

15. EXPECTED SUBMISSION DATE SYSTEM MONTH DAY COMPONENT MANU-FA CTURER REPORTABLE TO EPlX

EVENT DESCRIPTION LlCENS EE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

CONTINUATION SHEET On February 21, 2012, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was performing a normal shutdown in preparation for refueling outage 2R27. With Unit 2 at approximately 11.42%

power, the reactor was manually tripped in accordance with the 21/22/23 Feedwater f eater' Hi Hi alarm response procedure. There were no other unusual 1 not understood events associated with the shutdown.

1. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 2

EVENT ANALYSIS

Condensate is taken from the condenser2 hotwell by the condensate pumps3 and pumped through the filterldemineralizer system4 or its bypass line, the air ejector5 condensers, gland steam6 condenser, and low pressure (LP) heaters to the suction of the feedwater pumps7. The feedwater pumps then send feedwater through the high-pressure heaters to each steam generator.

2. DOCKET NUMBER 05000 306 The two main feedwater pumps operate in series with the condensate and the heater drain pumps, discharging through check valves and motor-operated gate valves into the No. 5 heaters. The feedwater flows through the two parallel, high-pressure feedwater heaters and flows into a common header. Two lines feed the two steam generators from the header.

Reheaters are provided with drain tanks and level controls. All the low-pressure feedwater heaters, No. 21, 22, and 23, are located in the condenser neck. Feedwater heaters No. 21 and 22 are combined into one shell (duplex) with bolted-head construction. Feedwater heater No. 21 is provided with a separate Feedwater Heater Drain Cooler.

6. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO 2012 -

001

- 00 The level controllers operate the emergency drain dump valves which dump the various drains directly to the condenser in case of abnormally high level. Three half-capacity, vertical, centrifugal heater drain pumps are provided for pumping the heater drainage into the condensate line ahead of the feedwater pumps. The pumps are started and stopped from the main control board. Tank level is controlled by variable-speed pump drives.

3. PAGE 2 OF 4 The sizing and operation of the Feedwater Heater (FWH) level control system at PINGP has not changed or degraded recently in a way that would have caused the Hi Hi feedwater levels. Further, no equipment malfunctions were identified. With no degraded or nonfunctioning components discovered and the sizing of the control valve and piping being the same as original design, the issue points to both an existing vulnerability in the system, and also recent operational changes.

' Ells System Code:

2 S B Ells System Code:

3 SG Ells Component Identifier: P Ells System Code:

SF Ells System Code:

6 SH Ells System Code:

TC Ells System Code:

SJ

The manual actuation of the Reactor protection system (RPS)' in accordance with the 21/22/23 Feedwater Heater Hi Hi alarm response procedure is reportable under 10 CFR 50.73(a)(2)(iv)(A). LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (1 0-201 0)

CONTINUATION SHEET

SAFETY SIGNIFICANCE

I.

FACILITY NAME Prairie Island Nuclear Generating Plant Unit 2 This event did not challenge nuclear safety as all plant systems responded as designed. Therefore, this event does not represent a safety system functional failure for Unit 2. However, the reactor was manually tripped in accordance with the alarm response procedure. After the reactor trip, Unit 2 continued the planned 2R27 shutdown to Mode 5. Therefore there were no radiological, environmental, or industrial impacts associated with this event and PlNGP did not affect the health and safety of the public.

CAUSE

2. DOCKET NUMBER 05000 306 The Equipment Cause Evaluation (ECE) determined that the third stage low pressure FWH bypass line to the condenser is potentially restricting flow to the dump valve. This by-pass line is presently two inches in diameter and supplies flow to a four inch control valve. Increasing this line to a larger diameter pipe will eliminate a restriction point for condensation to exit the third stage FWH.

The ECE also determined that PlNGP ramps the Moisture Separator Reheater Control valves closed earlier than necessary and are fully closed by approximately 20% reactor power. This causes excessive moisture in the extraction steam at low power and results in more water accumulating in the low pressure feedwater heaters.

6. LER NUMBER YEAR SEQUENTIAL REV NUMBER NO 2012 -

001

- 00

CORRECTIVE ACTIONS

3. PAGE 3 0 F 4

1. Start up and shut down procedures will be revised to preclude potential trip on Hi Hi low pressure feedwater heater level. This includes the associated Alarm Response Procedures and determining if the step to trip the turbine or reactor on Hi Hi LP FWH level is necessary.

2. An Equipment Improvement Long Range Plan Request (EIR) will be submitted to study increasing the size of the third stage FWH shell side bypass lines on both units to ensure they are not restricting flow to the condenser at low power operation.

3. The adverse trend with respect to Feedwater Heater Hi-Hi Level Alarms was identified as a programmatic weakness. Root Cause Evaluation (RCE) 01 326556-01 has been initiated to address the issue.

8 Ells System Code:

JC

PREVIOUS SIMILAR EVENTS LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION (1 0-201 0)

CONTINUATION SHEET A LER search was conducted and no similar events involving a manual reactor trip in accordance with an alarm response procedure were identified in the last three years at PINGP.

1. FACILITY NAME Prairie Island Nuclear Generating Plant Unit 2 However, the low pressure feedwater heater level indication reading Hi Hi at low power operations is a repeat event that led to a turbine trip on PINGP Unit 1 in April of 201 1 (Apparent Cause Evaluation (ACE) 012831 19). Corrective actions from this ACE included an Engineering Change (EC) to modify the Feedwater Heater Piping and a Condition Evaluation (CE) to evaluate revisions to shutdown operating procedures. An Equipment Improvement Long Range Plan Request (EIR) to obtain funding for a modification study of the LP FWH piping systems has been submitted.

Also, in June of 2005, while taking PINGP Unit 2 off line, immediately after opening the Unit 2 generator output breakers, the control room received a 21122123 Feedwater Heater Hi Hi Level Alarm.

The operators manually tripped the turbine per the alarm response procedure (Action Request (AR) 00855993). AR 00855993 was written to document the event and stated that this had happened previously in November, 2004 while performing the Turbine Overspeed Trip Exercise Surveillance Procedure (SP) 1036 for PINGP Unit 1. A CE determined that an engineering evaluation (EWR039553) was needed to evaluate the known deficiencies. No information could be found regarding this engineering evaluation.

2. DOCKET NUMBER 05000 306 This adverse trend with respect to Feedwater Heater Hi-Hi Level Alarms at PINGP was identified as a programmatic weakness. Root Cause Evaluation (RCE) 01326556-01 has been initiated to address the issue.

6. LER NUMBER SEQUENTIAL REV YEAR NUMBER NO 2012 -

001

- 00

3. PAGE 4 0 F 4